Abstract: Provided are a tube sheet of a steam generator installed to support a tube of the steam generator, and a manufacturing method thereof, particularly, a tube sheet of a steam generator having an anticorrosive layer and a manufacturing method thereof.

Abstract: A zirconium alloy for use in nuclear fuel assemblies is provided, which provides increased resistance against oxidation and corrosion and also improved bonding with parent material, because pure metallic material such as silicon (Si) or chromium (Cr) is evenly coated on the surface of the parent material by plasma spraying. Because the plasma spray coating used to coat the pure metallic material on the zirconium alloy does not require vacuum equipment and also is not limited due to the shape of the coated product, this is particularly useful when evenly treating the surface of the component such as 4 m-long tube or spacer grip arrangement which is very complicated in shape. Furthermore, because the coated zirconium alloy confers excellent resistance to oxidation and corrosion under emergency such as accident as well as normal service condition, both the economic and safety aspects of nuclear fuel are improved.

Abstract: A monitoring method of metal ions or oxygen ions applicable to a high concentration non-aqueous electrolyte includes: applying a potential in a non-aqueous electrolyte to obtain current information with respect to the potential; varying the potential applied in the non-aqueous electrolyte containing metal ion concentration or oxygen ion concentration such that the metal ion concentration or the oxygen ion concentration is maintained in spite of the potential being applied; detecting a linear relationship among the concentration, the current, and passed charges in the non-aqueous electrolyte by repeatedly performing the obtaining step and the varying step, while changing the concentration; and calculating metal ion concentration or oxygen ion concentration of the non-aqueous electrolyte in pyroprocessing of the non-aqueous electrolyte by using the linear relationship.

Abstract: Disclosed are a zirconium alloy for a nuclear fuel cladding having a good oxidation resistance in reactor accident conditions, a zirconium alloy nuclear fuel cladding prepared by using thereof and a method of preparing the same. The zirconium alloy includes 1.0 to 1.2 wt % of niobium (Nb); at least one element selected from tin (Sn), iron (Fe) and chromium (Cr); 0.02 to 0.1 wt % of copper (Cu); 0.1 to 0.15 wt % of oxygen (0); 0.008 to 0.012 wt % of silicon (Si) and a remaining amount of zirconium (Zr). The amount of Sn is 0.1 to 0.3 wt %, the amount of Fe is 0.3 to 0.8 wt %, and the amount of Cr is 0.1 to 0.3 wt %. A good oxidation resistance of the nuclear fuel cladding may be confirmed under accident conditions as well as normal operating conditions of a reactor, thereby improving economic efficiency and safety.

Abstract: Disclosed are a zirconium alloy for a nuclear fuel cladding having a good oxidation resistance in a severe reactor operation condition and a method of preparing zirconium alloy nuclear fuel claddings by using thereof. The zirconium alloy includes 1.8 to 2.0 wt % of niobium (Nb); at least one element selected from iron (Fe), chromium (Cr) and copper (Cu); 0.1 to 0.15 wt % of oxygen (O); 0.008 to 0.012 wt % of silicon (Si) and a remaining amount of zirconium (Zr). The amount of Fe is 0.1 to 0.4 wt %, the amount of Cr is 0.05 to 0.2 wt %, and the amount of Cu is 0.03 to 0.2 wt %. A good oxidation resistance of the nuclear fuel cladding may be confirmed under a severe reactor operation condition at an accident condition as well as a normal operating condition of a reactor, thereby improving economic efficiency and safety.

Abstract: Disclosed are a zirconium alloy for a nuclear fuel cladding having a good corrosion resistance by reducing an amount of alloying elements and a method of preparing a zirconium alloy nuclear fuel cladding using thereof. The zirconium alloy includes 0.2 to 0.5 wt % of niobium (Nb); 0.2 to 0.6 wt % of iron (Fe); 0.3 to 0.5 wt % of chromium (Cr); 0.1 to 0.15 wt % of oxygen (O); 0.008 to 0.012 wt % of silicon (Si) and a remaining amount of zirconium (Zr). The total amount of the niobium, the iron and the chromium is 1.1 to 1.2 wt %. A good oxidation resistance of the nuclear fuel cladding may be confirmed under accident conditions as well as normal operating conditions of a reactor, thereby improving economic feasibility and safety.

Abstract: A method for fabricating porous UO2 sintered pellets to be fed into the electrolytic reduction process for the purpose of metallic nuclear fuel recovery is provided, which includes forming a powder containing U3O8 by oxidizing spent nuclear fuel containing uranium dioxide (UO2) (step 1), fabricating green pellets by compacting the powder formed in step 1 (step 2), fabricating UO2+x sintered pellets by sintering the porous U3O8 green pellets fabricated in step 2 at 1200 to 1600° C., in an atmospheric gas (step 3), and forming UO2 sintered pellets by cooling the UO2+x sintered pellets to room temperature, and reduction the same at 1000 to 1400° C., in a reducing atmosphere (step 4).